Characterization of Microstructure and Property Evolution in Advanced Cladding and Duct: Materials Exposed to High Dose and Elevated Temperature

Designing materials for performance in high-radiation fields can be accelerated through a carefully chosen combination of advanced multiscale modeling paired with appropriate experimental validation. The studies reported in this work, the combined efforts of six universities working together as the Consortium on Cladding and Structural Materials, use that approach to focus on improving the scientific basis for the response of ferritic–martensitic steels to irradiation. A combination of modern modeling techniques with controlled experimentation has specifically focused on improving the understanding of radiation-induced segregation, precipitate formation and growth under radiation, the stability of oxide nanoclusters, and the development of dislocation networks under radiation. Experimental studies use both model and commercial alloys, irradiated with both ion beams and neutrons. Transmission electron microscopy and atom probe are combined with both first-principles and rate theory approaches to advance the understanding of ferritic–martensitic steels

A Systematic Study of Radiation-Induced Segregation in Ferritic–Martensitic Alloys

A systematic approach to measuring radiation-induced segregation (RIS) was used on four ferritic–martensitic (F–M) alloys: T91, HCM12A, HT9, and a Fe–9Cr model alloy, irradiated with 2.0 MeV protons over a range of doses (1–10 dpa) and temperatures (300–700°C). The experimental conditions are established so as to isolate the dependence of RIS on the experimental parameters: temperature, dose and bulk composition. RIS is measured at prior austenite grain boundaries (PAGBs) using the STEM/EDX technique. Chromium is found to enrich at PAGBs in all conditions with the exception being T91 irradiated to 3 dpa at 700°C. The magnitude of enrichment is small (\u3c2 at%). Minor elements Si, Ni, and Cu also enrich consistently. A bell-shaped temperature dependence of RIS is observed in all elements. The amount of Cr enrichment decreases as a function of increasing bulk Cr concentration. Lastly, it is found that the 9Cr model alloy reaches a steady-state Cr RIS behavior at approximately 7 dpa, while the T91 reaches what may be a steady state near 3 dpa, then the amount of enrichment decreases at 10 dpa

Irradiation-induced NanoCluster Evolution

Oxide dispersion strengthened steel (ODS) and commercial ferritic-martensitic (F-M) alloys are widely accepted candidate structural materials for designing advanced nuclear reactors. Nanoclusters embedded in the steel matrix are key microstructural features of both alloy types. Irradiation from nuclear fusion and fission affects the morphology of these nanoparticles, altering the performance of the alloys and potentially decreasing their usable lifetime. Thus, it is important to understand the effect of irradiation on these nanoparticles in order to predict long-term nuclear reactor performance. It was found that the evolution of nanoclusters in each material is different depending on the experimental irradiation parameters. The Nelson-Hudson-Mazey (NHM) model has been refined based on previous experimental work, and has been shown to be an effective model to simulate irradiation-induced nanocluster evolution in ODS and F-M steels. In this work, an NHM simulation tool was developed for nanoHUB, with a simplified user interface that enables rapid prediction of the effect of irradiation on the size of nanoclusters in a variety of Fe-based steels

Irradiation-Induced Amorphous-to-Crystalline Phase Transformations in Ceramic Materials

Amorphous ceramics are a unique class of materials with unusual properties and functionalities. While these materials are known to crystallize when subjected to thermal annealing, they have sometimes been observed to crystallize athermally when exposed to extreme irradiation environments. Because irradiation is almost universally understood to introduce disorder into materials, these observations of irradiation-induced ordering or crystallization are unusual and may partially explain the limited research into this phenomenon. However, the archival literature presents a growing body of evidence of these irradiation-induced amorphous-to-crystalline (a-to-c) phase transformations in ceramics. In this perspective, the summary and review of examples from the literature of irradiation-induced a-to-c transformations for various classifications of ceramics are provided. This work will highlight irradiation conditions and material parameters that appear most influential for activating a-to-c transformations, identify trends, examine possible mechanisms, and discuss the impact of a-to-c transformations on material properties. Finally, future research directions that will enable researchers to harness a-to-c transformations to tailor materials behaviors will be provided

Comparison of PM-HIP to Cast Alloy 625 for Nuclear Applications

PM-HIP, or Powder Metallurgy and Hot Isostatic Pressing, metals have been a low cost alternative to forged and cast structural metals within various industries. The nuclear industry has recently developed interest in PM-HIP alloys, but further research needs to be done to quantify their mechanical properties and characterize the microstructure. Specifically, we must understand the mechanical and microstructural evolution of PM-HIP materials after long-term operation at the elevated temperatures that PM-HIP components will experience in service. We focus on Ni-base alloy Inconel 625, and compare the PM-HIP version to the cast version. Our methodology consists of annealing samples to various temperatures, 400,600, and 800 °C, at various times, 100, 1,000, and 10,000 hours, to see the temperature and time effect on these alloys. We conduct microhardness testing and optical microscopy to evaluate the strength and grain size, respectively. We have found that average grain size in PM-HIP 625 samples are consistently smaller than that in cast 625, and this grain size difference persists with heat treatment. Future work will involve scanning electron microscopy (SEM) imaging and tensile testing of the annealed specimens, as well as irradiation exposure

Enhanced radiation damage tolerance of amorphous interphase and grain boundary complexions in Cu-Ta

Amorphous interfacial complexions are particularly resistant to radiation damage and have been primarily studied in alloys with good glass-forming ability, yet recent reports suggest that these features can form even in immiscible alloys such as Cu-Ta under irradiation. In this study, the mechanisms of damage production and annihilation due to primary knock-on atom collisions are investigated for amorphous interphase and grain boundaries in a Cu-Ta alloy using atomistic simulations. Amorphous complexions, in particular amorphous interphase complexions that separate Cu and Ta grains, result in less residual defect damage than their ordered counterparts. Stemming from the nanophase chemical separation in this alloy, the amorphous complexions exhibit a highly heterogeneous distribution of atomic excess volume, as compared to a good glass former like Cu-Zr. Complexion thickness, a tunable structural descriptor, plays a vital role in damage resistance. Thicker interfacial films are more damage-tolerant because they alter the defect production rate due to differences in intrinsic displacement threshold energies during the collision cascade. Overall, the findings of this work highlight the importance of interfacial engineering in enhancing the properties of materials operating in radiation-prone environments and the promise of amorphous complexions as particularly radiation damage-tolerant microstructural features

Comparison of PM-HIP to Forged SA508 Pressure Vessel Steel Under High-Dose Neutron Irradiation

Powder metallurgy with hot isostatic pressing (PM-HIP) is an advanced manufacturing process that is envisioned to replace forging for heavy nuclear components, including the reactor pressure vessel (RPV). But PM-HIP products must at least demonstrate comparable irradiation tolerance than forgings in order to be qualified for nuclear applications. The objective of this study is to directly compare PM-HIP to forged SA508 Grade 3 Class 1 low-alloy RPV steel at two neutron irradiation conditions: ~0.5-1.0 displacements per atom (dpa) at ~270C and ~370C. PM-HIP SA508 experiences greater irradiation hardening and embrittlement (total elongation) than forged SA508. However, uniform elongation and approximate toughness are comparable across all irradiated materials, suggesting irradiated PM-HIP SA508 exhibits superior ductility at maximum load-bearing capacity. The irradiation hardening mechanism is linked to composition rather than fabrication method. Since PM-HIP SA508 has higher Mn and Ni concentration, it is more susceptible to irradiation-induced nucleation of Mn-Ni-Si-P (MNSP) nanoprecipitates and dislocation loops, which both contribute to hardening. Conversely, the forged material nucleates fewer MNSPs, causing dislocation loops to control irradiation hardening. These results show promise for the irradiation performance of PM-HIP SA508 and can motivate future nuclear code qualification of PM-HIP fabrication for RPVs

Corrigendum to Comparing Structure-Property Evolution for PM-HIP and Forged Alloy 625 Irradiated with Neutrons to 1dpa [Mater. Sci. Eng. A (2022) 144058]

The authors regret that after publication, they discovered that the dislocation loop number density was undercounted by a factor of 100 for both the PM-HIP and forged specimens. While this does not change the original major conclusions, this necessitates a change in the results presentation (Sections 3.2 and 4.1) and calculated hardening (Table 3, Fig. 5). Corrections to these affected sections are provided in this corrigendum

Nanoindentation Investigation of Chloride-Induced Stress Corrosion Crack Propagation in an Austenitic Stainless Steel Weld

Transgranular chloride-induced stress corrosion cracking (TGCISCC) is a mounting concern for the safety and longevity of arc welds on austenitic stainless steel (AuSS) nuclear waste storage canisters. Recent studies have shown the key role of crystallography in the susceptibility and propagation of TGCISCC in SS weldments. Given that crystallography underlies mechanical heterogeneities, the mechanical-crystallographic relationship during TGCISCC growth must be understood. In this study, welded SS 304L coupons are loaded in four-point bend fixtures and then boiled in magnesium chloride to initiate TGCISCC. Nanoindentation mapping is paired with scanning electron microscopy (SEM) electron backscatter diffraction (EBSD) to understand the correlation between grain orientation, grain boundaries, and hardening from TGCISCC propagation. The nanoindentation hardness of individual grains is found to not be a controlling factor for TGCISCC propagation. However, intragranular hardness is generally highest immediately around the crack due to localized strain hardening at the crack tip. This work shows that nanoindentation techniques can be useful in understanding CISCC behaviors when paired with electron microscopy